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in subjects: "Years Living Radioisotopes 050800* -- Nuclear Fuels-- Spent Fuels Reprocessing"

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Actinide partitioning-transmutation program final report. I. Overall assessment

This report is concerned with an overall assessment of the feasibility of and incentives for partitioning (recovering) long-lived nuclides from fuel reprocessing and fuel refabrication plant radioactive wastes and transmuting them to shorter-lived or stable nuclides by neutron irradiation. The principal class of nuclides considered is the actinides, although a brief analysis is given of the partitioning and transmutation (P-T) of /sup 99/Tc and /sup 129/I. The results obtained in this program p… more

Date: June 1, 1980

Creator: Croff, A. G.; Blomeke, J. O. & Finney, B. C.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Analysis of Lime-Slurry Stirred Tank Carbonation Reactor

Gas residence time distributions were determined for a stirred tank carbonation reactor. Empirical correlations for the first and second moments of the residence time distribution (RTD) curves as functions of flow rates and impeller speeds were obtained. Decontamination factors for /sup 85/Kr were measured.

Date: September 23, 1977

Creator: McAleese, J. P.; Belt, B. A.; Datesh, J. R. & Shaeffer, M. C.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Application of on-line plutonium isotopic concentration monitors at a nuclear fuel reprocessing plant

Gamma-ray spectroscopy is used to assay plutonium solutions flowing through thin sample cells. The direct measurement of gamma-ray intensities are used to determine the plutonium isotopic content and total concentration. The development of spectral analysis algorithms, the use of appropriate instrument internal standards, techniques to minimize spectral acquisition times, and methods to deduce the /sup 242/Pu content are discussed. Results of experiments on plutonium solutions in the 200 to 500… more

Date: February 1, 1979

Creator: Hofstetter, K. J. & Huff, G. A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Calculated critical parameters in simple geometries for oxide and nitrate water mixtures of U-233, U-235 and Pu-239 with thorium. Final report

Calculations have been performed on water mixtures of oxides and nitrates of /sup 233/U, /sup 235/U, and /sup 239/Pu with chemically similar thorium compounds to determine critical dimensions for simple geometries (sphere, cylinder, and slab). Uranium enrichments calculated were 100%, 20%, 10%, and 5%; plutonium calculations assumed 100% /sup 239/Pu. Thorium to uranium or plutonium weight ratios (Th: U or Pu) calculated were 0, 1, 4, and 8. Both bare and full water reflection conditions were ca… more

Date: November 1, 1979

Creator: Converse, W.E. & Bierman, S.R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Characterization of airborne plutonium-bearing particles from a nuclear reprocessing plant

The elemental compositions, sizes, structures, and /sup 239/Pu contents were determined for 558 plutonium-bearing particles isolated from airborne particles collected at various locations in the exhaust from a nuclear fuel reprocessing facility. These data were compared with data from natural aerosol particles. Most of the collected particles were composed of aggregates of crustal materials. 3.6% of the particles were organic and 1.7% were metallic, viz., iron, chromium, and nickel. High enrich… more

Date: January 1, 1978

Creator: Sanders, S. M.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Comparative studies of head-end processing using irradiated, mechanically blended and coprecipitated (U,Pu)O/sub 2/ reactor fuels

A series of experiments was performed with highly irradiated (U,Pu)O/sub 2/ to determine the effects of voloxidation on the release of fission products and on the subsequent dissolution characteristics. It was found that voloxidation was more effective at 600 than at 650/sup 0/C for releasing the gaseous fission products /sup 3/H, /sup 14/C, and /sup 85/Kr from the experimental fuels. The lower-temperature treatment released about 95% of the tritium, 97 to 98% of the /sup 14/C, and about 56% of… more

Date: September 1, 1978

Creator: Goode, J. H. & Stacy, R. G.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Comparison of experimentally determined spent-fuel compositions with ORIGEN 2 calculations

The specific experimental measurements of interest here involve the determination of parameters related to the actinide and fission product composition of samples from five elements taken from fuel assemblies discharged from the Turkey Point Unit 3 PWR. Two fuel assemblies were obtained for the purposes of nondestructive and destructive assay. These assemblies were initially fueled with 448 kg of UO/sub 2/ enriched to 2.556 wt % /sup 235/U and then irradiated for 851 full-power days. Five eleme… more

Date: January 1, 1981

Creator: Croff, A. G.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Comparison of silver sorbents for application to radioiodine control at the PUREX process facility modification. [Iodine 129]

In continued support of the design of the gaseous radioiodine control system for the PUREX Process Facility Modification (PFM), the Pacific Northwest Laboratory (PNL) conducted laboratory-scale measurements of the performance of four state-of-the-art sorbents for radioiodine in the dissolver offgas (DOG) of a nuclear reprocessing plant. The PFM is a new head-end treatment plant being designed by Westinghouse Hanford Company (WHC) for the PUREX Plant at the Hanford Site. The experiments performe… more

Date: September 1, 1988

Creator: Scheele, R.D.; Burger, L.L. & Halko, B.T.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Computer-controlled on-line gamma analysis for krypton-85

/sup 85/Kr will be evolved from spent nuclear fuel during both the voloxidation and dissolution processes, so a reliable method for on-line analysis of /sup 85/Kr in the off-gas system is needed. Tritium, /sup 14/C, and /sup 129/I were trapped, and the activity of /sup 85/Kr was then measured using a Li-drifted Ge detector. Equipment used to carry out this analysis is described; the PET computer is used. The /sup 85/Kr evolution rate was correlated with the fuel dissolution rate; the close corr… more

Date: March 1, 1980

Creator: Canuette, R.P.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Consolidated fuel reprocessing program. Developments for the future in reprocessing

The future reprocessing developments focus on three major areas: (1) the retention of gaseous fission products to reduce off-site doses to very low values; (2) the initial steps of breakdown, shearing, and dissolution of breeder fuels; and (3) advanced facility and equipment concepts, which are expected to lead to a reliable, cost-effective, totally remotely operated and maintained plant. Work in the first area - removal of fission gases (the most important of which is /sup 85/Kr) - is largely … more

Date: January 1, 1982

Creator: Burch, W.D.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Containment and recovery system for fuel-reprocessing plants

Tritium containment and removal problems in a fuel-reprocessing plant are identified and conceptual process designs for reducing emissions to the environment to below 1 Ci/day are studied. The conceptual design recommended would allow an air atmosphere in the reprocessing-plant hall and would use a continuous-catalytic-oxidizer/molecular-sieve-adsorber cleanup system to maintain a 40-..mu..Ci/m/sup 3/ tritium level (5 ..mu..Ci/m/sup 3/ HTO) against 180 Ci/day leakage from components and process… more

Date: August 25, 1976

Creator: Galloway, T. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Coordination chemistry of several radius-sensitive complexones and applications to lanthanide-actinide separations

The relationships between the lanthanide complex formation equilibria and the lanthanide-actinide separation application of three radius sensitive ligands have been studied. The consecutive stepwise formation constants of the 1:1, 2:1, and 3:1 chelate species formed by the interaction of DHDMB and the tripositive lanthanides and yttrium were determined potentiometrically at 0.1 M ionic strength and 25/sup 0/C. Results indicate that three different coordination modes, one tridentate and two bide… more

Date: October 1, 1981

Creator: Potter, M.W.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Criticality experiments with low enriched UO/sub 2/ fuel rods in water containing dissolved gadolinium

The results obtained in a criticality experiments program performed for British Nuclear Fuels, Ltd. (BNFL) under contract with the United States Department of Energy (USDOE) are presented in this report along with a complete description of the experiments. The experiments involved low enriched UO/sub 2/ and PuO/sub 2/-UO/sub 2/ fuel rods in water containing dissolved gadolinium, and are in direct support of BNFL plans to use soluble compounds of the neutron poison gadolinium as a primary critic… more

Date: February 1, 1984

Creator: Bierman, S. R.; Murphy, E. S.; Clayton, E. D. & Keay, R. T.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Determination of tritium in solutions from nuclear fuel reprocessing

A method for determining the tritium content of solutions containing high levels of radioactivity was developed. The solutions which contain tritium as HTO are decontaminated by deionization with a mixture of cation and anion exchange resins (H and OH forms). The HTO concentration in the resulting deionized water is then determined by liquid scintillation counting. The method has been applied to light-water reactor fuel dissolver solutions and to a simulated dissolver solution containing known … more

Date: January 1, 1977

Creator: Baumann, E. W. & MacMurdo, K. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Development of Ag/sup 0/Z for bulk /sup 129/I removal from nuclear fuel reprocessing plants and PbX for /sup 129/I storage

Tests were conducted to develop Ag-exchanged mordenite (AgZ) for removal of gaseous /sup 129/I from nuclear fuel reprocessing plants. The effects of bed depth and hydrogen pretreatment on the elemental (I/sub 2/) iodine loading of AgZ were examined. The tests indicated that reduced AgZ (Ag/sup 0/Z) had about twice the capacity for iodine as AgZ, and at least 15-cm bed depths should be used for loading tests. The effects of H/sub 2/O(g), NO, NO/sub 2/, and bed temperature on the iodine loading o… more

Date: January 1, 1978

Creator: Thomas, T. R.; Staples, B. A. & Murphy, L. P.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Developments in effluent control for breeder reprocessing facilities

10 CFR 20 and 40 CFR 190 impose tight restrictions on the releases of radioactive materials from the fuel cycle. Retention factors needed in the LMFBR and LWR fuel cycles to meet the 40 CFR 190 release limits of /sup 85/Kr, /sup 129/I, and /sup 239/Pu are listed. Technologies for the control of I, /sup 85/Kr, and tritium are discussed. Advanced effluent control systems will be demonstrated in the Hot Experimental Facility. (DLC)

Date: January 1, 1980

Creator: Yarbro, O.O. & Sears, M.B.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Distribution of thorium during full loading of carboxylic acid cation exchange resins with uranium from nitrate solutions at 30 and 40/sup 0/C

The recycle of /sup 233/U for the refabrication of fuel elements for a high-temperature gas-cooled reactor will require an assessment of chemical specifications for impurities in the uranyl nitrate product stream from the fuel reprocessing facility. The behavior of thorium, as an impurity in this simulated process stream, was investigated under equilibrium conditions during full loading of carboxylic acid cation exchange resins with uranium. Mathematical correlations of the data indicate a comp… more

Date: December 1, 1978

Creator: Shaffer, J. H. & Greene, C. W.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Electrolytic dissolver for plutonium metal

Laboratory tests have demonstrated that plutonium metal, as well as aluminum, stainless steel, Hastelloy C, and other alloys, can be electrolytically dissolved in nitric acid using the ''solution contact'' approach. Two sources of plutonium metal were evaluated, weapons' grade and a special reactor grade (42% /sup 240/Pu, 12% /sup 241/Pu) from Savannah River, both yielding a dissolution rate of approximately one gram per ampere-hour. A current flow of 150 amperes was attained at 12 volts (limit… more

Date: September 20, 1971

Creator: Fox, R. D.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Evaluation of target power supplies for krypton storage in sputter-deposited metals

Implantation of /sup 85/Kr in a growing sputtered metal deposit has been studied for the containment of /sup 85/Kr recovered from the reprocessing of spent nuclear fuel. PNL, as part of DOE's research program for /sup 85/Kr storage, has developed krypton trapping storage devices (KTSDs) in a range of sizes for ''cold'' and radioactive testing. The KTSD is a stainless steel canister that contains a sputtering target for depositing an amorphous rare-earth transition metal on the inner wall and si… more

Date: April 1, 1986

Creator: Greenwell, E. N.; McClanahan, E. D. & Moss, R. W.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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The evolution of iodine during metal dissolution

During the dissolution in nitric acid of uranium metal which has been irradiated in a pile, radio-iodine (8.0 day) and radio-xenon (5.3 day) are liberated along with the oxides of nitrogen. Since in the plant these gases are discharged into the atmosphere, the presence of the activities (particularly radio-iodine because of its greater chemical reactivity) presents a physiological hazard. As a consequence, the cooling period (that is, the time between the discharge of the metal from the pile an… more

Date: August 2, 1945

Item Type: Report

Partner: UNT Libraries Government Documents Department

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High temperature gas-cooled reactor criticality research program

Objective is to define the nuclear criticality research program needed in support of the HTGR fuel cycle outside reactors. An experimental plan and supportive calculational efforts are presented. Currently, some data on well-moderated /sup 235/U and /sup 233/U exist which are applicable to thorium--uranium fuel mixtures. However, experimental data are needed for undermoderated /sup 235/U and /sup 233/U systems. Data are needed also on /sup 235/U and /sup 233/U containing soluble neutron poisons… more

Date: June 1, 1976

Creator: Bierman, S. R. & Clayton, E. D.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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High-temperature gas-cooled reactor fuel recycle development. Annual progress report for period ending September 30, 1977

The status of the following tasks is reported: program management, studies and analysis, fuel processing, refabrication development, in-plant waste treatment, research general support, and major facilities including HTGR recycle reference facility, hot engineering test facility and cold prototype test facility-refabrication. (JRD)

Date: September 1, 1978

Creator: Lotts, A. L. & Kasten, P. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Hot cell studies of light water reactor fuel reprocessing. [Radioactivity distribution; dissolution; extraction; waste evaporation]

Experiments were run using all-glass equipment with irradiated fuel in a hot cell, in order to study possible problems in LWR fuel reprocessing (radioactivity distribution during dissolution, stability of extraction raffinate, waste partitioning). Behavior of /sup 129/I, tritium, and /sup 14/C in the off-gas was studied, as were the radioisotopes in the dissolver solution; results are compared with ORIGEN calculations. Actinide mass analyses were also made and compared with ORIGEN. Residues and… more

Date: January 1, 1976

Creator: Campbell, D. O. & Buxton, S. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Hot cell studies of tritium removal from and dissolution of an irradiated thoria fuel

Tritium removal and dissolution studies on an irradiated thoria-based fuel were conducted in the High Level Caves at the Savannah River Laboratory (SRL). The objectives of these studies were to define the effects of key process-related parameters on tritium evolution and subsequent dissolution. The test program at SRL determined the effects on tritium removal of particle size, heating temperature, oxidation, and agitation. ThO/sub 2//UO/sub 2/ (95%/5%) fuel from the Elk River Reactor, irradiate… more

Date: January 1, 1980

Creator: Pickett, J.B.

Item Type: Article

Partner: UNT Libraries Government Documents Department